Table of contents

In three-dimensional configurations, the confinement region is surrounded by the stochastic magnetic field lines related to magnetic islands or separatrix, leading to the fact that the plasma–vacuum boundary is not so definite compared with tokamaks that the various modulations of the plasma–vacuum boundary will be induced around the stochastic region by synergetic effects between a transport around the stochastic region and a large Shafranov shift of the whole plasma, in especially high-β operations. To examine such modulation effects of the plasma boundary on MHD instabilities, high-β plasmas allowing a large Shafranov shift or a large Pfirsch–Schlüter current are considered in the inward-shifted LHD configurations with the vacuum magnetic axis R_ax of 3.6 m, for which previous theoretical analyses based on fixed MHD equilibria indicate that pressure-driven modes are significantly more unstable compared with experimental observations. The concept of the averaged flux surfaces allowing a movement of the equilibrium plasma into the stochastic region is introduced, which induces a boundary modulation and, at the same time, reduces the discrepancy on MHD equilibria between the experimentally obtained and theoretically considered. As a result, it is shown that the boundary modulation, namely, the whole plasma outward-shift due to a large Pfirsch–Schlüter current has significant stabilizing effects on ideal MHD instabilities, leading to partially resolving the discrepancy on MHD stability between experimental results and theoretical analyses.

In the Tohoku University Heliac (TU-Heliac), a helical axis stellarator, an electron injection electrode using a hot cathode made of LaB₆ was developed and the transition mechanism to an improved mode has been intensively studied. In the electrode current-sweep mode of biasing experiments, a bifurcation phenomenon, i.e. a negative resistance feature in the electrode characteristics was observed accompanied with transition to an improved mode (H-mode) or transition from H-mode to L-mode, in helium plasma discharges of a wide range of collisionality. The ion viscous damping force was estimated from the J × B driving force for poloidal rotation. The local maxima in viscosity were found at the poloidal Mach number around −M_p ∼ 1–3, as predicted by neoclassical theory. It was also found that the negative resistance (a bifurcation phenomenon) was observed when the poloidal viscosity showed a local maximum.

Plasma current start-up to 100 kA was achieved successfully in the JT-60U tokamak without the use of the centre solenoid (completely solenoidless tokamak operation). Only poloidal field coils located on the outboard side of the torus were used, in combination with strong ionization by electron cyclotron (EC) power. The presence of a field null was not necessary for plasma current start-up, but the flux conversion efficiency was low in such a case. In a nearly solenoidless start-up, low neutral pressures were favoured, and the optimum location of the EC resonance was slightly to the high field side of the vacuum vessel centre. The required EC power for efficient utilization of flux swing in JT-60U was about 1 MW. A plasma current of 260 kA was maintained for 1 s by NB only, and plasma current ramp-up from 215 to 310 kA was achieved by EC and neutral beam (NB) only (without lower hybrid current drive (LHCD)). However, the ramp-up efficiency was much lower compared with LHCD. Recharging of the centre solenoid was observed with only counter and perpendicular NB injection, indicating bootstrap overdrive. Integration of these elements can lead to the achievement of a completely solenoidless tokamak operation.

Recent research on advanced tokamak in JET has focused on scenarios with both monotonic and reversed shear q-profiles having plasma parameters as relevant as possible for extrapolation to ITER. Wide internal transport barriers (ITBs), r/a ∼ 0.7, are formed at ITER relevant triangularity δ ∼ 0.45 and moderate plasma current, I_P = 1.5–2.5 MA, with n_e/n_G ∼ 60% when ELMs are moderated by Ne injection. At higher current (I_P ≤ 3.5 MA, δ ∼ 0.25) wide ITBs sitting at r/a ≥ 0.5, in the positive shear region, have been developed. Generally MHD events terminate these barriers otherwise limited in strength by power availability. ITBs with core density close to Greenwald value, T_e ∼ T_i and low toroidal rotation (4 times lower than standard ITBs) are obtained in plasma target preformed by opportune timing of lower hybrid current drive (LHCD), pellet injection and a small amount of NBI power. Wide ITBs, r/a ∼ 0.6, of moderate strength, can be sustained without impurities accumulation for a time close to neoclassical resistive time in 3 T/1.8 MA discharges that exhibit reversed magnetic shear profiles and type-III ELMy edge. These discharges have been extended to the maximum duration allowed by JET subsystems (20 s) bringing to the record of injected energy in a JET discharge: E ∼ 330 MJ. Portability of ITB physics has been addressed through dedicated similarity experiments. The ITB is identified as a layer of reduced diffusivity studying the propagation of the heat wave generated by modulating the ICRF mode conversion (MC) electron heating. Impressive results, Q_DT ∼ 0.25, are obtained in these deuterium discharges with ³He minority when the MC layer is located in the core. The ion behaviour has been investigated in pure LHCD electron ITBs optimizing the ³He minority concentration for direct ion heating. Preliminary results of particle transport, studied via injection of a trace of tritium and an Ar–Ne mixture, will be presented.

The problem of producing fusion power with low neutron emission has been debated in the past in the framework of the magnetic confinement fusion research. Proposals are still being renewed to use advanced fuels in various plasma systems. Since today's toroidal devices cannot support plasma conditions suitable for a large fusion production with such fuels, new concepts and configurations have been studied, where the plasma components are not in a thermal equilibrium. Here, a system of a neutral beam injected into a confined plasma is considered where fusion is produced only between the beam and plasma ions. The collisional slowing down of the beam into the plasma is described by a fluid model. General considerations in this model allow conditions to be found for the fusion-produced power to breakeven against the power needed to sustain the system itself. These conditions are only necessary since the nuclear power is maximized in the present analysis by using favourable assumptions. Nevertheless, the results for different advanced fuels indicate again the very high difficulty of getting a net power produced by the fusion of such fuels, unless the plasma target temperature reaches very high and unrealistic values.

High confinement and stability are obtained simultaneously in stationary conditions in improved H-mode discharges at ASDEX Upgrade. The improved H-mode discharges are typically composed of two different phases: 'lower heating phase', where H₉₈(y, 2) is similar to standard H-modes (H₉₈(y, 2) ∼ 1), and 'fully developed improved H-mode phase', where H₉₈(y, 2) is higher than standard H-modes (H₉₈(y, 2) up to 1.4). In this paper, the confinement physics is studied by comparing these two different phases using ASTRA simulations. The results are compared with experimental observations. Firstly, the time evolution of the q-profile is simulated with ASTRA using experimental kinetic profiles and compared with experimental measurements. Secondly, the two different phases are compared by power balance analyses and predictive modelling using the Weiland model with ASTRA. Ion temperature gradient lengths and turbulence levels measured by reflectometry are compared between the two phases. Lastly, the roles of density peaking and pedestal pressure in confinement improvement are discussed. The transport analyses using the ASTRA code and experimental observations show that the fully developed improved H-mode phase is very similar to the lower heating phase in terms of core heat transport. With respect to the confinement improvement, it is suggested that the increase of the edge pedestal pressure plays an important role. Enhancement of the radial electric field at the edge is believed to be closely linked with confinement improvement at the edge pedestal region in the fully developed phase.

Neutron emission spectroscopy diagnostic has been used to study plasmas produced at JET during the main deuterium–tritium experimental campaign (DTE1). The plasmas were subjected to radio frequency (RF) and neutral beam (NB) heating in different combinations during periods of individual discharges. Similar discharges with different heating combinations were also compared. This paper deals with the possibility of observing direct evidence of synergetic coupling of the ion cyclotron resonance frequency of the RF power to the fast ions from the NB injection. The key signature of such an effect is the high-energy tail in the spectrum of the neutron emission from plasmas subjected to either NB or RF or both simultaneously. Four discharges of the DTE1 campaign were identified as most suitable to search for the synergetic effects including NB(d) and NB(t) heated discharges with RF coupled to the second harmonic resonance of d and t ions. The measured spectra were analysed with up to four components and interpreted in terms of underlying velocity components of the ions responsible for the neutron emission. The results of the analysis are presented and studied from the aspect of finding effects of the application of NB and RF power injection, which go beyond being merely superpositional. Enhanced effects of simultaneous injection are found which are ascribed as synergetic. The strength of the synergetic effect is found to vary depending on the plasma parameters at hand, but this needs further studies to shed light on the underlying physics of the synergetic NB + RF heating effects on the plasma.

Initial results demonstrating the concept of constant inductive helicity injection are presented. Constant helicity injection is achieved using two oscillating inductive helicity injectors, with the goal of producing a bow tie spheromak. Each injector is a 180° segment of a reverse field pinch and they are driven 90° out of phase. Approximately 5 MW of power is injected during the 6 ms pulse, and the input power has been maintained at a fairly constant value by directly fuelling the injectors with neutral gas. Motivation for the experiment is given, including beta-limit calculations for the bow tie spheromak. Fuelling the injectors with neutral gas during the discharge is shown to produce injector parameters that are more constant in time. A series of discharges with increasing power input shows a promising increase in toroidal current. Unique construction techniques of the experiment are also described.

The evolution of recycling behaviour has been investigated during long pulse discharges in the HT-7 tokamak after ICRF boronization (C₂B₁₀H₁₂) using the H/(H+D) ratio and the edge recycling coefficient R. After boronization, impurity reduction is observed, attributed to the fresh boron film, but the recycling coefficient can exceed unity due to a large amount of hydrogen absorbed in the film, leading to an uncontrollable density rise and discharge termination. When the H/(H+D) ratio was reduced to less than 25%, the electron density was easily controlled. The longest discharge, up to 240 s with central electron temperature T_e(0) of about 1.0 keV and central electron density n_e(0) of 0.8 × 10¹⁹ m⁻³, was achieved following boronization. After many discharges the effectiveness of boron film was weakened, and the density rise was correlated with an increase in both carbon and oxygen radiation which limited the duration of long pulse discharges.

A key feature of the 'high recycling steady' (HRS) H-mode regimes is the reduction of the transient heat load to the divertor target due to large ELMs. The heat flux to the divertor targets measured by means of a Langmuir probe in the scrape-off layer on the target plate is about ∼0.3 MW m⁻² in the typical HRS plasmas, which is very small in comparison with the transient heat load in the ELMy H-mode discharge. In addition, the effects of the plasma shaping on the access to and improved plasma performance of HRS H-mode regimes have been investigated. Furthermore, simultaneous measurements of both density fluctuations and potential profiles at the plasma edge region using a heavy ion beam probe are also presented.

Extensive analysis of disruptions in JET has helped advance the understanding of trends of disruption-generated runaway electrons. Tomographic reconstruction of the soft x-ray emission has made possible a detailed observation of the magnetic flux geometry evolution during disruptions. With the aid of soft and hard x-ray diagnostics runaway electrons have been detected at the very beginning of disruptions. A study of runaway electron parameters has shown that an approximate upper bound for the conversion efficiency of pre-disruptive plasma currents into runaways is about 60% over a wide range of plasma currents in JET. Runaway generation has been simulated with a test particle model in order to verify the results of experimental data analysis and to obtain the background for extrapolation of the existing results onto larger devices such as ITER. It was found that close agreement between the modelling results and experimental data could be achieved if in the calculations the post-disruption plasma electron temperature was assumed equal to 10 eV and if the plasma column geometry evolution is taken into account in calculations. The experimental trends and numerical simulations show that runaway electrons are a critical issue for ITER and, therefore, the development of mitigation methods, which suppress runaway generation, is an essential task.

This paper presents the significant progress made in the research and development (R&D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li₂TiO₃ pebble bed under neutron-pulsed operation conditions are summarized.

With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150 °C followed by normalizing it at 930 °C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R&D on the breeder material, Li₂TiO₃, the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li₂TiO₃ pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14 MeV neutrons with an accuracy of 10%. The tritium release rate from the Li₂TiO₃ pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation.

The Large Helical Device (LHD) achieves high-performance plasma confinement by the coordination of the magnetic surface region and the chaotic field line layer. It is theoretically and experimentally shown that drift surfaces exist for highly energetic particles being extended over the last closed flux surface (LCFS) in the LHD. These particles are considered lost particles due to the loss-cone in the previous theories, where the analyses are limited inside the LCFS. The present theory predicts that the loss-cone is strongly reduced in the LHD and that highly energetic particles confined over the LCFS exist. These are consistent with the LHD experimental results in both the ICRF heating experiments and the low magnetic field neutral beam injection heating experiments. From particle orbit analyses and studies on the connection length of diverter field lines, it is also shown that plasma can exist in the chaotic field line layer located outside the LCFS in the LHD. The plasma in the chaotic field line layer is clearly detected by CCD-cameras in the LHD experiment. This ambient plasma might be expected to play the role of a kind of impregnable barrier for the core plasma, which suppresses both the MHD instabilities and the cooling of the core plasma due to charge exchange processes. The line-tying effects of diverter field lines that are slipped out from the chaotic field line layer can also stabilize the ballooning mode and the vertical displacement events of the plasma column.

Steady state fully non-inductive current plasmas in Tore Supra offer an opportunity to study the turbulent particle pinch. The local parametric dependence of the particle pinch velocity allows us to discriminate between different theories. The existence of an anomalous pinch is unambiguously demonstrated. The results support the turbulent theories based on ion temperature gradient and trapped electron modes. An anomalous inward pinch related to the safety factor (q) profile is found to be dominant in the gradient region (normalized radius 0.3 ⩽ r/a ⩽ 0.6). In contrast, the direction of the anomalous pinch in the plasma core (r/a ⩽ 0.3) is more sensitive to the electron temperature (T_e) profile.

In the present study, fluctuation characteristics have been investigated using beam emission spectroscopy in discharges with the edge transport barrier in the compact helical system. If the heating power was high enough, a coherent mode having fundamental (f = 4–5 kHz) and 2nd harmonic (2f = 8–10 kHz) frequencies was observed. This mode showed similar characteristics to those of the edge harmonic oscillations (EHO) observed in tokamaks. This EHO-like mode is located in the edge region where steep density gradients and the ι = 1 rational surface exist. There seems to be a threshold in the density gradient for enhancing the EHO-like mode. After the excitation of the EHO-like mode, the density gradient saturates. The edge density gradient remained constant when the EHO-like mode existed, even if the heating power was increased, suggesting that there is some correlation between the EHO-like mode and particle confinement.

Fast-ion instabilities with frequencies somewhat below the ion cyclotron frequency occur frequently in spherical tokamaks such as the National Spherical Torus Experiment (NSTX). NSTX and the DIII-D tokamak are nearly ideal for fast-ion similarity experiments, having similar neutral beams, fast-ion to Alfvén speed v_f/v_A, fast-ion pressure, and shape of the plasma but with a factor of two difference in major radius. When DIII-D is operated at low field (0.6 T), compressional Alfvén eigenmode (CAE) instabilities appear that closely resemble the NSTX instabilities. In particular, the mode frequencies, polarization and beam-energy threshold are nearly identical to NSTX. CAE in high-field discharges and emission at cyclotron harmonics are also observed. As on NSTX, the basic stability properties are consistent with the idea that the instability is driven by anisotropy in the fast-ion velocity distribution and is damped predominantly by Landau damping of electrons. The results suggest that these modes might be excited in ITER.

Typical equilibrium reconstructions for axisymmetric plasma confinement devices involve a nonlinear least squares minimization with respect to parameters of the Grad–Shafranov equation. These parameters specify the toroidal field and pressure as functions of the poloidal flux, and the fit minimizes a cost function, χ², which involves measurements such as the tangential poloidal field near the boundary. Most fitting procedures in use weigh the terms in χ² by the variances of the measurements, but none accounts for the possibility of correlations between the measurements. Results presented here indicate that the presence of correlations may have a strong effect on the fidelity of the fit, possibly beneficial when they are optimally accounted for but quite detrimental otherwise. These results also show that increasing the number of measurements of the same type can have little effect when the additional measurements are within a correlation scale length of the previous measurements. A further issue is that the toroidal field and pressure terms in the Grad–Shafranov equation play nearly the same role, leading to a near-degeneracy in the fit. Internal pressure measurements most effectively mitigate this near-degeneracy when correlations are optimally treated.

In the period 1998—2001 the JET tokamak was operated with the MkII Gas Box divertor. On two occasions during that period a number of limiter and divertor tiles were retrieved from the torus and then examined ex situ with surface sensitive techniques. Erosion and deposition patterns were determined in order to assess the material erosion, material migration and fuel inventory on plasma facing components. Tracer techniques, e.g. injection of ¹³C labelled methane and tiles coated with a low-Z and high-Z marker layer, were used to enhance the volume of information on the material transport. The results show significant asymmetry in the distribution of fuel and plasma impurity species between the inner (net deposition area) and the outer (net erosion) divertor channels. No significant formation of highly hydrogenated carbon films has been found in the Gas Box structure. The important processes for material migration, and the influence of operation scenarios on the morphology of the deposits are discussed. Comparison is also made with results obtained following previous divertor campaigns.

The injection of high-pressure supersonic jets into the tokamak plasma is considered a promising method of future thermonuclear reactor fuelling and as a tool for disruption mitigation. Successful experiments were correspondingly performed on Tore Supra and DIII-D. In the present paper the evolution of such a jet is analysed. The jet expansion, deceleration of the ambient electrons and ions by the jet, self-consistent electric field, elementary processes, radiation and adiabatic cooling of the ambient plasma are taken into account. The jet is simulated by a MHD code, which is similar to the code previously used for pellets. It is demonstrated that the ionization degree of the jet strongly depends on the jet parameters. Several simulations were performed for the range of parameters typical for DIII-D. The jet of initial density 4 × 10²⁴ m⁻³ remains almost neutral, and only the outer regions are ionized. When the initial jet density is reduced by a factor of 2 or more the main part of the jet becomes ionized rather fast. It is demonstrated that ionization at the jet edge in the poloidal (perpendicular to the magnetic field) direction of the jet is sufficient to stop poloidal expansion of the jet by force. The final poloidal size of the jet remains of the order of its initial poloidal dimension (of the order of ten centimetres). The jet motion in the radial direction (direction of the injection) is provided by the polarization poloidal electric field and the corresponding drift. In the paper two mechanisms of polarization reduction are considered: Alfvén conductivity of the ambient plasma and the ∇B-induced drift. It is shown that an almost neutral jet can penetrate deep into the tokamak while a modest ionization degree should prevent its penetration for the case of low field side injection.

The critical beta, which is limited by an external kink mode with Alfvén wave time scale, is shown to be decreased by ferromagnetic effect by about 8% for μ/μ₀ ∼ 2 where μ and μ₀ denote the permeability of a ferromagnetic wall and vacuum, respectively, for a tokamak of aspect ratio 3. The existence of the stability window in the wall distance for resistive wall mode opened by the effects of both the toroidal plasma rotation and the plasma dissipation, which was not observed for the high aspect ratio tokamak, is found for the tokamak of aspect ratio 3. The effect of ferromagnetism on them is also investigated. The critical beta analyses of the National Centralized Tokamak (NCT) plasma using the VALEN code are started by stabilizing plate and vacuum vessel geometry with finite resistivity, and the results of the passive effect of the stabilizing plate are obtained. The calculations including the stabilizing effect of the vacuum vessel and also active feedback control are performed for a design of the NCT plasma.

This paper reports on ITER-relevant ion cyclotron resonance frequency (ICRF) physics investigated on JET in 2003 and early 2004. Minority heating of helium three in hydrogen plasmas—(³He)H—was systematically explored by varying the ³He concentration and the toroidal phasing of the antenna arrays. The best heating performance (a maximum electron temperature of 6.2 keV with 5 MW of ICRF power) was obtained with a preferential wave launch in the direction of the plasma current. A clear experimental demonstration was made of the sharp and reproducible transition to the mode conversion heating regime when the ³He concentration increased above ∼2%. In the latter regime the best heating performance (a maximum electron temperature of 8 keV with 5 MW of ICRF power) was achieved with dipole array phasing, i.e. a symmetric antenna power spectrum. Minority heating of deuterium in hydrogen plasmas—(D)H—was also investigated but was found inaccessible because this scenario is too sensitive to impurity ions with Z/A = 1/2 such as C⁶⁺, small amounts of which directly lead into the mode conversion regime. Minority heating of up to 3% of tritium in deuterium plasmas was systematically investigated during the JET trace tritium experimental campaign (TTE). This required operating JET at its highest possible magnetic field (3.9 to 4 T) and the ICRF system at its lowest frequency (23 MHz). The interest of this scenario for ICRF heating at these low concentrations and its efficiency at boosting the suprathermal neutron yield were confirmed, and the measured neutron and gammay ray spectra permit interesting comparisons with advanced ICRF code simulations. Investigations of finite Larmor radius effects on the RF-induced high-energy tails during second harmonic (ω = 2ω_c) heating of a hydrogen minority in D plasmas clearly demonstrated a strong decrease in the RF diffusion coefficient at proton energies ∼ 1 MeV, in agreement with theoretical expectations. Fast wave heating and current drive experiments in deuterium plasmas showed effective direct electron heating with dipole phasing of the antennas, but only small changes of the central plasma current density were observed with the directive phasings, in particular at low single pass damping. New investigations of the heating efficiency of ICRF antennas confirmed its strong dependence on the parallel wavenumber spectrum. Advances in topics of a more technological nature are also summarized: ELM studies using fast RF measurements, the successful experimental demonstration of a new ELM-tolerant antenna matching scheme and technical enhancements planned on the JET ICRF system for 2006, they being equally strongly driven by the preparation for ITER.